1) JSC “OKBM Afrikantov”, Nizhny Novgorod, Russia 2) FSUE “SSC RF-IPPE named after A. I. Leypunsky” , Obninsk, Russia

5th Joint IAEA-GIF Technical Meeting/Workshop on

Safety of Sodium-Cooled Fast Reactors

Alexey V. Vasyaev1), Sergey F. Shepelev1), Igor A. Bylov1)

Vladimir М. Poplavskiy2), Iurii M. Ashurko2)

IAEA HQ, Vienna , Austria, 23 - 24 June 2015

Provision of

Safety Design Criteria

for Generation-IV

Sodium-cooled Fast Reactor System

implementation

in BN-1200 reactor design

BN-1200 design

Safety design criteria for Generation-IV sodium-cooled fast reactor

system

BN-1200:

Safety criteria

Safety concept

Engineered safety provisions

Safety analysis and assessment

Conclusion

2

Contents

The BN-1200 next generation sodium-cooled fast reactor is developed in accordance with Federal Target Programme "A New Generation of Nuclear Power Technologies for 2010-2015 and Prospectively until 2020"

The BN-1200 is next generation reactor plant which should meet all Generation IV systems goals in:

Sustainability

Safety

Economics

Proliferation resistance and physical protection

BN-1200 reactor is considered as sodium-cooled fast reactor design for serial construction

3

BN-1200 design

1 - IHX; 2 – main vessel; 3 – guard vessel; 4 – supporting structure; 5 – inlet plenum; 6 – core debris tray; 7 - core; 8 – pressure pipeline; 9 - MCP-1; 10 – refueling mechanism; 11 - CRDM; 12 – rotating plugs.

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Parameter Value

Thermal power, МW 2800

Electric power, MW 1220

Capacity factor, % 90

Efficiency, %:

brutto netto

43.5 40.7

Cooling circuits number 3

Primary and secondary circuits coolant Sodium

3rd circuit coolant Water/steam

Desing lifetime, years 60

Sodium temperature in the primary circuit at core inlet/outlet, °C

410/550

Sodium temperature in the secondary circuit at SG inlet/outlet, °C

355/527

Live steam pressure, MPa

Live steam temperature, °C Feedwater temperature, С

17.0

510

275

Key BN-1200 parameters

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Safety Design Criteria for

Generation-IV Sodium-cooled Fast Reactor System

Developed by the GIF Safety Design Criteria

Task Force in 2013

Include:

Safety approach to the SFR as a Generation-IV

reactor system

Management of safety in design

Principal technical criteria

General plant design

Design of specific plant systems

SDC analysis indicates that:

Criteria correspond to the IAEA Specific Safety

Requirements SSR-2/1 Safety of Nuclear Power

Plants: Design

Criteria correspond to the Russian national

regulatory NPP safety requirements (OPB-88/97,

NP-082-07) in accordance with BN-1200 design is

developed

Radiation safety: allowable radiation doses at nuclear power plant site:

Normal operation: less 20 mkSv per year at and beyond site boundary

Anticipated operational occurrences: 100 mkSv per year at and beyond site

boundary

Design-basis accidents: not greater than 1 mSv for first year after accident at

site boundary excluding any protective measures beyond site boundary

Beyond design-basis accidents: not greater than 50 mGy beyond site

boundary at early accident stage (first 10 days) and not greater than 50 mSv for

first year after accident (population evacuation and resettlement are excluded)

Probabilistic safety goals:

Cumulative severe accident probability for one nuclear plant unit for one

year should not exceed 10-6. Severe accident is beyond design-basis accident

with fuel damage exceeding maximum design margin

Cumulative large early release probability for one nuclear plant unit for

one year should not exceed 10-7. Large early release is a release of

radioactive materials to environment during accident at NPP when public

protection measures beyond emergency planning zone boundary are required

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Safety criteria

BN-600

1980

BR-5/10

1959

50 years of successful design and operating experience

BN-800 reactor commissioning at final stage

Opportunity for commercial fast reactor construction: BN-1200

BOR-60

1969

BN-1200

Development

BN-350

1973

BN-800

Commissioning

Power reactors Experimental

reactors

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Proven engineering practices

BN-1200 reactor safety is ensured through consistently

implementing the defense-in-depth concept based on:

Physical barriers in confining radioactive material at specified

locations

System of technical and organizational measures aimed at

preventing harmful effects of radiation on people and the

environment, and ensuring adequate protection from harmful

effects and mitigation of the consequences in the event that

prevention fails

Physical barriers include:

Fuel composition

Fuel cladding

Primary circuit boundary: reactor vessel, intermediate heat

exchangers tubes, piping

Plant containment

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Safety concept

9

Inherent safety features (I)

Characteristic Effect produced

Large heat capacity of integral reactor design

(all primary circuit equipment contained inside

the reactor vessel)

Low heat-up rate of sodium in the primary

circuit (20°C an hour) in case of complete

loss of cooling after emergency shutdown

Placement of all systems containing

radioactive sodium within the reactor vessel

Exclusion of radioactive sodium release into

rooms from external piping

Low corrosive and erosive effect produced by

sodium upon structural materials

Exclusion of fuel rod overheating as a

consequence of clogging of the fuel assembly

open flow area and breach of reactor vessel

as a result of corrosion

High boiling temperature of sodium: 883°С at

atmospheric pressure

Low, approximately 0.15 MPa, absolute

pressure in the reactor gas cavity (sodium

boiling temperature 930°C); absence of

coolant phase transitions in case of primary

circuit depressurization with reliable core

cooling

Characteristic Effect produced

Effective containment of iodine and caesium

isotopes by sodium (confirmed by BN-600

operational experience)

Practically full preclusion of any release of

the most hazardous gaseous and air-borne

fission product isotopes into the environment

during normal operation and accidents

Stable negative reactivity coefficient for

reactor power (dK/dN = -2,6*10-6 MW-1)

and core materials temperature

(dK/dT = -2,3*10-5 °С-1)

Limited reactor power increase in case of

unauthorised introduction of positive

reactivity, power drop in case of

unauthorised reactor heat-up

Introduction of negative reactivity in the

event of sodium boiling as a consequence of

presence of the top sodium cavity in the fuel

assemblies

Reactor power drop in case of core

overheating accident to sodium boiling

temperature

Design of control and protection rods drive

system that precludes simultaneous

withdrawal of more than one rod from the

core

Exclusion of unauthorised insertion of

excessive positive reactivity and dangerous

reactor power excursion

Inherent safety features (II)

In addition to the inherent safety features BN-1200 design includes

number of passive safety devices and passive safety systems

Hydraulically suspended absorber rods

Absorber rods with temperature-triggered action

Decay heat removal system with air heat exchangers

Leak-tight above-reactor premise

Guard vessel

Core catcher of a heat-resistant metal at the bottom of the reactor vessel

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Passive safety systems

Design basis impacts:

maximum design-basis earthquake: magnitude 7 on the MSK-64

scale

design-basis earthquake: magnitude 6 on the MSK-64 scale

20 tons aircraft crash on the reactor building

external pressure wave with front pressure 30 kPa

Beyond design basis impacts:

400 tons commercial aircraft crash on the reactor building

considering fuel ignition

earthquake with horizontal ground acceleration 40% more than

maximum design-basis earthquake

12

External hazards consideration

To practically exclude radioactive sodium leaks into reactor rooms and accordingly prevent any sodium fires associated with generation of air-borne radioactivity following engineering solutions are adopted in BN-1200 design :

Placement of all systems containing radioactive sodium within the reactor vessel

Guard vessel

Exclusion of the spent fuel assemblies drum – a potential source of accidents involving radioactive sodium

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Design engineering solutions to increase safety

14

Design engineering solutions to increase safety:

measures to minimize possibilities of non-radioactive sodium leaks (1)

BN-800 SG

BN-1200 SG

Evaporator

Superheater

Sodium inlet collector

Buffer tank

Applying of shell-type steam generators (8 modules) instead of module-type

steam generators BN-600 (72 modules) and BN-800 (60 modules) substantial

decreasing of secondary circuit pipes length

Air heat

exchanger

Secondary circuit

main pump

Reactor

Interim heat

exchanger

Expansion

bellows

SG

Design engineering solutions to increase safety:

measures to minimize possibilities of non-radioactive sodium leaks (2)

Effective prevention and

sodium leaks consequences

mitigation

Applying of built-up pipeline

shrouds sodium leaks

localization

Applying of expansion bellows

thermal displacement

compensation

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To assess safety deterministic and probabilistic safety analyses shall

be done for all operational conditions

To assess BN-1200 safety all technically credible events even with low

probability are taken into consideration

Following postulated initiating events are taking into account:

Main reactor vessel leakage: Decrease of coolant level inside reactor

caused by leakage into space between main and guard vessel shall not

lead to break of sodium circulation in primary circuit

Steam generator leakage: The total cross section rupture of one steam

generator tube is considered as postulated initiating event. Additional

damage of tubes surrounding ruptured tube can be considered as beyond

design-basis event

Total instantaneous blockage of a fuel subassembly: Sodium boiling

and fuel rods damage (cladding and fuel melting) in one fuel assembly

without propagation of damage in the remaining volume of the core

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Safety analysis and assessment (I)

In addition to regulatory requirements in BN-1200 safety analysis

following events are considered:

Failure of all active safety systems

Single failures in passive safety systems

The most severe beyond design basis accidents are analyzed in

BN-1200 design as following:

Station blackout and reactor shutdown systems failure

Station blackout and emergency heat removal system failure

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Safety analysis and assessment (II)

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Conclusions

BN-1200 reactor design provides Safety Design Criteria for

Generation-IV Sodium-cooled Fast Reactor System implementation

Sodium-cooled fast reactors design and operating experience

accumulated in Russia confirms technology maturity and ability for

developing BN-1200 as sodium-cooled fast reactor design for

serial construction

BN-1200 reactor design safety provisions due to optimal

combination of proven and new solutions allow to consider this design

as design of Generation IV systems

Thank you for

attention!